Retention or release of radionuclides in a deep geological repository for radioactive wastes strongly depends on the geochemical environment and on the interaction with near-field components, e.g. waste packages and backfill materials. Deep geological disposal in rock salt is one of the concepts considered for cemented low- and intermediate-level wastes. Long-term experiments were performed to observe the evolution of full-scale cemented waste simulates (doped with (NH4)(2)U2O7) upon reaction with relevant salt brines, e.g. MgCl2-rich and saturated NaCl solutions, and to examine the binding mechanisms of uranium. Throughout the experiments, concentrations of major solution components, uranium and pH values were monitored regularly and compared to thermodynamic equilibrium calculations, which indicate that close-to-equilibrium conditions have been achieved after 13-14 years duration of the leaching experiments. Two of the full-scale cemented waste simulates were recovered from the solutions after 17-18 years and studied by different analytical methods to characterize the solids, especially with respect to uranium incorporation. In drill core fragments of various lateral and horizontal positions of the corroded monoliths, U-rich aggregates were detected and analyzed by means of space-resolved techniques. Raman, mu-XANES and mu-XRD analyses of several aggregates demonstrate that they consist of an amorphous diuranate-type solid. Within error, calculated U solubilities controlled by Na-diuranate (Na2U2O7.H2O) are consistent with measured U concentrations in both, the NaCl and the MgCl2-system. Since uranophane occurs also in the corroded monoliths, it is proposed that a transition towards the thermodynamic equilibrium U(VI) phase is kinetically hindered. (C) 2013 Elsevier Ltd. All rights reserved.

Low- and intermediate-level radioactive wastes are frequently solidified in a cement matrix. In a potential repository for nuclear wastes, the cementitious matrix is altered upon contact with solution and the resulting secondary phases may provide for significant retention of the radionuclides incorporated in the wastes. In order to assess the secondary phases formed upon corrosion in chloride-rich solutions, which are relevant for nuclear waste disposal in rock salt, leaching experiments were performed. Conventional laboratory batch experiments using powdered hardened cement paste in MgCl2-rich solutions were left to equilibrate for up to three years and full-scale cemented waste products were exposed to NaCl-rich and MgCl2-rich solutions for more than twenty years, respectively. Solid phase analyses revealed that corrosion of hardened cement in MgCl2-rich solutions advanced faster than in NaCl-rich solutions due to the extensive exchange of Mg from solution against Ca from the cementitious solid. Thermodynamic equilibrium simulations compared well to results at the final stages of the respective experiments indicating that close to equilibrium conditions were reached. At high cement product to brine ratios (>0.65 g mL−1), the solution composition in the laboratory-scale experiments was close to that of the full-scale experiments (cement to brine ratio of 2.5 g mL−1) in the MgCl2 systems. The present study demonstrates the applicability of thermodynamic methods used in this approach to adequately describe full-scale long-term experiments with cemented waste simulates.

Cementation is an industrial scale conditioning method applied to fix and solidify liquid low and intermediate level radioactive wastes (LLW/ILW) prior to underground disposal in geological formations.To assist prognosis of the long-term safety of cemented waste, alteration of uranium doped cement productswas studied in chloride-rich solutions relevant for final LLW/ILW disposal in rock salt. After long-time exposure of the full-scale LLW/ILW simulates to concentrated NaCl and MgCl2 brines, solid samples were retrieved for chemical and mineralogical analysis with an emphasis on uranium speciation in the corroded cement matrix.Bulk and recent spatially resolved micro(μ) U L3-XAFS measurements point to the occurrence of a diuranate type U(VI) phase forming throughout the corroded cement monoliths. U-enriched hot spots with dimensions up to several tens of μm turn out to be generally X-ray amorphous.

Source terms are derived for different scenarios of a potential final repository in the salt dome Gorleben, under the precondition that NaCl- or MgCl2-rich solutions are present. Kinetic and thermodynamic mobilization / retention and the influence of temperatures are discussed and the expected geochemical conditions are analyzed. Upper limit concentrations for the radionuclides Am, Th, U, Np, Pu, Tc, Zr and rare earth elements are derived for the simplified scenarios Umfang: XII, 92 S.

In accordance with the Belgian “supercontainer design”, spent nuclear fuel (SNF) will be encapsulated in carbon steel canisters, surrounded by a concrete overpack for disposal in poorly-indurated clay. After re-saturation of the barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Corrosion studies of SNF in ECW-type solution (Evolved Cement Water) and YCWCa-type solution (Young Cement Water with Ca) were performed under externally applied H2 overpressures over 426 days. Directly after H2 application, Tc concentrations decreased from >10−8 M to concentrations below detection limit. Based on the fractional release of selected fission products, low matrix dissolution rates of ~10−8/day were found in both experiments. U concentrations decreased finally to 1.5•10−9 M (YCWCa) and to 2.1•10−10 M (ECW), respectively. Am, Np and Pu concentrations were found throughout the experiments below their detection limits indicating an effective retention process.

Even though chemical processes related to the corrosion of spent nuclear fuel in a deep geological repository are of complex nature, knowledge on underlying mechanisms has very much improved over the last years. As a major result of numerous studies it turns out that alteration of irradiated fuel is significantly inhibited under the strongly reducing conditions induced by container corrosion and consecutive H-2 production. In contrast to earlier results, radiolysis driven fuel corrosion and oxidative dissolution appears to be less relevant for most repository concepts. The protective hydrogen effect on corrosion of irradiated fuel has been evidenced in many experiments. Still, open questions remain related to the exact mechanism and the impact of potentially interfering naturally occurring groundwater trace components. Container corrosion products are known to offer considerable reactive surface area in addition to engineered buffer and backfill material. In combination, waste form, container corrosion products and backfill material represent strong barriers for radionuclide retention and retardation and thus attenuate radionuclide release from the repository near-field.

In pulse radiolysis experiments at 22°C production and decay of Cl2− radicals are studied in 0.1 and 1.0mol dm−3 NaCl solutions by observing its optical absorption over a time span of 20ms. In addition to experiments with NaCl solutions equilibrated with N2O at ambient pressure, a series of experiments are conducted with NaCl solutions equilibrated at 10MPa hydrogen partial pressure and 0.1MPa N2O partial pressure. In the presence of hydrogen, the Cl2− yield is significantly reduced compared to that in hydrogen free experiments. The effect of hydrogen on the radiolytic yield is more pronounced in 0.1mol dm−3 NaCl solution than in the relatively concentrated NaCl solution. In parallel to the experiments the evolution of the Cl2− concentration is simulated using a kinetic model. Based on the comparison between measured and simulated optical absorption, which is mainly caused by the Cl2− radical, the rate constant of reaction Cl2−+Cl2−=Cl−+Cl3− is determined as 5.2(±0.8)×108Lmol−1s−1 at zero ionic strength. This value is within the range of published rate constants for the Cl2− disproportionation reaction.

Radionuclide source terms are derived for different disposal options of a potential final repository in the salt dome Gorleben. The geochemical conditions are analyzed under the precondition that NaCl, MgCl2- or CaCl2-rich solutions are present. Kinetic and thermodynamic mobilization/retention and the influence of temperatures are discussed. Upper limit concentrations for the radionuclides Am, Th, U, Np, Pu, Tc, Zr and rare earth elements are derived for the simplified scenarios.